ML20205E325
| ML20205E325 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/19/1987 |
| From: | Wiggins J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20205E251 | List: |
| References | |
| 50-293-87-03, 50-293-87-3, NUDOCS 8703300592 | |
| Download: ML20205E325 (22) | |
See also: IR 05000293/1987003
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.
50-293
Report No.
87-03
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Licensee:
Boston Edison Company
800 Boylston Street
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Boston, Massachusetts 02199
Facility Name: Pilgrim Nuclear Power Station
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Inspection At: Plymouth, Massachusetts
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Inspection Conducted:
January 1, 1987 - February 20, 1987
Inspectors:
M. McBride, Senior Resident Inspector
J. Lyash, Resident Inspector
T. Kim, Resident Inspector
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L. Doerflien, Project Engineer
Approved By:
hE AJM[A
3h ff2
J. {idgins, gf, Reactor Projects Section 18
~ Date
Inspection Summary:
Areas Inspected:
Routine resident inspection of plant operations, radiation
protection, physical security, plant events, maintenance, surveillance, outage
activities, and reports to the NRC. Licensee preparations for reactor defuel-
ing and subsequent fuel movement activities were also reviewed.
Results: Three violations were identified concerning failure to properly con-
trol high radiation area keys, failure to initiate a Failure and Malfunction
Report, and failure to comply with fire protection system technical specifica-
tions.
Additional inspector concerns included the following:
The possible existence of a single failure affecting the standby gas
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treatment system which had not been identified during recent engineering
reviews (section 3.a).
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The use of an out of date procedure during conduct of a surveillance test
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(section 3.c).
8703300592 870323
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InspectionSummary(Continued)
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The absense of technical specifications for the RPS electrical protection
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assemblies (section 3.c).
The apparent failure of operations and fire protection personnel to
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recognize the status of fire protection equipment (section 3.d).
Poor contamination control and RWP procedure adherence practices (section
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3.e).
Submittal of an inaccurate Licensee Event Report (section 5).
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The inspectors also noted that the licensee's response to IE Information Notice
86-106, concerning piping erosion-corrosion appears aggressive and thorough.
Licensee preparation for and execution of fuel offload activities were well
organized and performed.
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TABLE OF CONTENTS
Pa9e
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1.
S umma ry o f Fa c i l i ty Ac t i v i t i e s . . . . . . . . . . . . . . . . . . . . . .
1
2.
Followup on Previous Inspection Findings ............
1
Violations, Unresolved Items and Inspector Follow Item
3.
Routine Periodic Inspections ........................
4
a.
System Alignment Inspection
b.
Plant Maintenance and Outage Activities
c.
Surveillance Testing
d.
Fire Protection
e.
Radiation Prot'ection
4.
Review of Plant Events ..............................
13
a.
Licensee Ma6agement and Organizational Changes
b.
c.
Reactor Defueling Preparations and Conduct
5.
Review of Licensee Event Reports (LERs) . . . . . . . . . . . . .
17
6.
Management Meetings .................................
18
Attachment I - Persons Contacted
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DETAILS
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1.0 Summary of Facility Activities
The plant was shutdown on April 12, 1986 for unscheduled maintenance. On
July 25,1986, Boston Edison announced that. the outage would be extended
to include refueling and completion of certain modifications.
A management conference was held at .the licensee's Chiltonville Training
Center' on February 2,1987. NRC and licensee senior management discussed
the progress of program improvements.
During the period, Rear Admiral
R.
G. Bird, U. S. Navy (Retired) was
appointed Senior Vice President-Naclear. On February 4, 1987 the licensee
announced the replacement of the plant manager, and several other station
organization changes.
The licensee commenced reactor defueling on the evening of February 6,
,
1987. The core was completely off-loaded by February 13, 1987.
A third full time NRC Resident Inspector was assigned to Pilgrim on
January 26, 1987.
2.0 Followup on Previous Inspection Findings
Violations
(Update) Violation (84-36-03), failure to continuously monitor the SRMs
during refueling. The inspector reviewed station procedure 1.3.34 Conduct
of Operations, Rev.12, and associated OPER 38, Shift Turnover Checklist.
The inspector also observed Nuclear Operation Supervisor and Nuclear Plant
Operator - shift turnovers and determined that adequate transmittal of
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information regarding plant status and changes in plant conditions were
performed.
This item remains open pending implementation of licensee
commitments, including Operations Section Manager or
Chief Operating
Engineer presence in the control room during important aspects of the
upcoming reactor restart.
(Closed) Violation (85-19-04), failure to specify surveillance frequencies
on Radiation Work Permit (RWPs) for high radiation areas. The inspector
reviewed revised station procedure 6.1.022, Revision 20, Issue, Use, and
Termination of Radiation Work Permits (RWP's). Surveillance frequencies
appear to be adequately addressed. The procedure requires the radiation
protection supervisor or designee to specify the surveillance frequency on
the RWP. It further states that the frequency shall be in specific terms
(i.e.,
once per shift, constant, every two hours, etc.).
The licensee
also instructed the health physics supervisors responsible for approving
RWP's that surveillance frequencies be specifically stated in the remarks
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section of the RWP.
The inspector reviewed selected RWP's during this
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inspection period and verified that the frequencies were specified. This
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item is closed.
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Unresolved Items
(Closed) Unresolved Item (86-25-10), review fire barrier discrepancies and
compliance with. technt. cal specifications.
A number of deficient fire
barriers.were identified by the licensee during April,1985. No specific
corrective actions were taken in response to these deficiencies until
July, 1986. The licensee's failure to effectively utilize the corrective
action program in addressing these items was identified as noncompliance
50-293/86-25-09.
In addition, the inspector questioned licensee compli-
ance with technical specifications during tSis period of time.
In re-
sponse, the licensee conducted a review of each discrepancy. This review,
described in the licensee's written response to inspection report 50-293/
86-25, indicated that fire watch coverage was provided in these areas
during the entire time period. This coverage, although not tied to the
specific deficiencies, provided compensatory measures equivalent to tech-
nical specification requirements. The inspector sampled licensee records
supporting this assertion.
No problems were identified.
This item is
closed.
(Closed) Unresolved Item (86-29-07), evaluate the reportability of fire
barrier deficiencies.
Failure and Malfunction Report number 86-164 iden-
tified nineteen potentially degraded fire barriers.
The inspector ques-
tioned the reportability of this condition.
Engineering evaluations per-
formed justify the operability of the existing barriers. This justifica-
tion was reviewed and approved by the licensee engineering department
under engineering disposition document SUDDS86-146.
Based upon documen'
ted analysis demonstrating that the barriers are capable of performing
their design function, this issue does not appear to be reportable.
(Closed) Unresolved Item (86-34-04), followup on recent fire barrier
inspections.
The licensee recently identified a large number of fire
barrier penetrations which were degraded, with inadequate qualification
documentation or for which proper surveillance was not performed.
In
addition, compensatory measures established by the licensee in response
to the number of deficiencies did not appear aggressive.
The failure of
the licensee to ensure proper implementation of compensatory fire watches
is the subject of noncompliance 50-293/86-36-04. The licensee has insti-
tuted an extensive program to identify all fire barrier penetrations and
upgrade appropriate surveillance procedures.
Review of progress in this
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area is tracked by unresolved item 50-293/86-36-02.
Based on the above
referenced open items, this item is considered administratively closed.
(Update) Unresolved Item (86-36-01), review licensee actions to implement
repairs on fire protection equipment and reduce the number of compensatory
fire watches. Similar concerns were identified as inspector follow items
50-293/86-38-03 and 86-06-12.
Items 86-38-03 and 86-06-12 have been
administratively closed based on the existence of item 86-36-01.
Future
inspection in this area should consider the licensee's commitments and
corrective actions contained in written responses to inspection reports
86-38 and 86-06.
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Inspector Follow Items
(Closed) Inspector Follow Item (83-06-01), review ifcensee actions to
assure other fire barrier penetrations are not missed from the program.
This item was last updated in inspection report 50-293/86-21. The licen-
see identified numerous fire barrier penetrations not included in appli-
cable surveillance procedures.
It was also noted that resources were not
available to properly maintain barrier surveillance procedures.
The
licensee has dramatically increased management oversight and staffing
levels in the area of fire protection.
An extensive effort to identify
and document all penetrations is underway.
The progress and completion
of this walkdown program is the subject of unresolved item 50-293/
86-36-02.
Based on ' the increase in fire protection staffing and the
existence of the above referenced item, this item is considered closed.
(Closed) Inspector' Followup Item (84-23-03), general plant working condi-
tions to be reviewed during routine safety inspections as followup re-
sponse to allegation made to OSHA involving drywell industrial safety
hazards. The licensee's records indicate that the item was resolved among
the licensee, General Electric Company, and Bechtel Power Company follow-
ing a meeting with an OSHA representative on July 27, 1984. The inspector
independently conducted tours of the drywell and other plant areas per-
iodically to observe plant safety conditions.
The general plant condi-
tions appeared satisfactory. The inspectors will continue to review plant
safety conditions during routine inspections. This item is closed.
(Closed) Inspector Follow Item (86-06-12), follow licensee efforts to
reduce the number of station fire watches. The licensee continues to rely
heavily on compensatory fire watches.
Regional specialist inspection
50-2f 3/86-36 identified similar concerns. Followup on continuing licensee
efforts to reduce the number of fire watches will be conducted under unre-
solved item 50-293/86-36-01.
Based on the above, this item is adminis-
tratively closed.
(Update) Inspector Follow Item (86-29-03), review licensee evaluation of
SBGT system single failures.
This item was last updated in inspection
report 50-293/86-37.
The inspectors questioned licensee engineering
personnel regarding a single failure of a backdraft damper which could
affect system operability.
This is described in section 3.a of this
report.
(Update) Inspector Follow Item (86-37-09), review licensee evaluation of
the seismic qualification of HGA relays. On January 16, 1987, the licen-
see reported via ENS that a potentially significant application of this
unqualified relay type had been identified. This is discussed in section
3.c of this report.
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(Closed) Inspector Follow Item (86-38-03), fire protection system mainten-
ance is not always properly prioritized. Similar observations were made
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during a specialist inspection and are documented as unresolved item
50-293/86-36-01.
Based on the existence of the above item, this item is
administratively closed.
3.0 Routine Periodic Inspections
The inspectors routinely toured the facility to assess general plant' and
equipment conditions, housekeeping and adherence to fire protection,
security and radiological control measures. Ongoing work activities were
monitored to verify that they were being conducted in accordance with
approved administrative and technical procedures, and that proper communi-
cations with the control room staff had been established. The inspector
observed valve, instrument and electrical equipment lineups in the field
to ensure that they were consistent with system operability requirements
and operating procedures.
During tours of the control room the inspectors verified proper staffing,
access control and operator attentiveness.
Adherence to procedures and
limiting conditions for operations were evaluated. The inspectors exam-
ined equipment lineup and operability, instrument traces and status of
control room annunciators. Various control rooms logs and other available
licensee documentation were reviewed.
In addition to routine equipment operability confirmation, the inspectors
performed independent walkdowns. of selected safety systems. Confirmation
of the as-built system configuration, identification of any degraded con-
ditions and procedure adequacy were evaluated.
The inspector observed and reviewed outage activities, maintenance and
problem investigation activities to verify compliance with regulations,
procedures, codes and standards.
Involvement of QA/QC, safety tag use,
personnel qualifications, fire protection precautions, retest require-
ments, and reportability were assessed.
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The -inspector observed tests to verify performance in accordance with
approved procedures and LCO's, collection of valid test results, removal
and restoration of equipment, and deficiency review and resolution.
Radiological controls were observed on a routine basis during the report-
ing period. Standard industry radiological work practices, conformance to
radiological control procedures and 10 CFR Part 20 requirements were
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observed.
Independent surveys of radiological boundaries and random
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surveys of radiologically clear points throughout the facility were taken
by the inspector.
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Checks were made to determine whether security conditions met regulatory
requirements, the physical security plan, and approved procedures. Those
checks included security staffing, protected and vital area barriers,
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personnel
identification,
access control,
badging,
and compensatory
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measures when required.
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a.
System Alignment Inspection
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Several potential single failures have been identified by the
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licensee which could affect the operability of the Standby Gas
Treatment Sys' tem (SBGT). These single failures have previously
been reported to the NRC under Part 21, and are the subject of
existing open item 50-293/86-29-03. Each failure identified by
the licensee depends on the presence of an air operated fail-
open damper which crossties the redundant SBGT system trains.
After completion 'of the licensee's engineering evaluation the
' inspectors independently examined the system configuration for
the existence of unidentified single failures.
The SBGT system consists of two treatment trains and two exhaust
fans with a common discharge duct. The trains are also connec-
ted at the outlet of the treatment trains, just upstream of the
exhaust fans, by an air operated fail-open crosstie damper.
A
backdraft damper is installed on the discharge of each fan to
prevent any recirculation flow. During the design basis acci-
dent the crosstie damper would fail open and both exhaust fans
would auto start. After a preset time delay one fan automati-
cally shuts down.
If the backdraft damper on the outlet of the
idle fan failed to fully close, a recirculation path for the
operating fan would be established. This condition could result
in a decrease in flow from the reactor building and result in a
loss of secondary containment. The licensee stated that a low
pressure sensor in the system common discharge duct should
activate to restart the idle fan if this failure were to occur.
Justification that this action would occur under all partial or
full failures of the backdraft damper could not be provided by
the licensee.
The inspector will review the licensee's final
analysis under existing open item 50-293/86-29-03. The accepta-
bility of the SBGT during fuel offload activities is discussed
in section 4.c of this report.
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The inspector reviewed the measures taken by the licensee to
ensure that systems susceptible to freezing were adequately pro-
tected.
The only safety related systems which are not buried or
in a heated area and therefore susceptible to freezing are the
main stack sample line and the jockey fire pump low pressure
switch sensing line.
Both lines are heat traced to prevent
freezing and the licensee verified proper operation of the heat
tracing circuits during November 1986.
In addition to these
annual circuit checks, the inspector noted other controls are in
place, such as weekly surveillance test 8.B.1, Fire Pump Test
which periodically verifies the heat tracing circuits are ener-
gized.
The inspector also noted that the auxiliary boilers,
which provide heating to the station's buildings and compart-
ments, receive extensive annual preventive maintenance during
the summer to ensure proper operation of the station's heating
system during cold weather.
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During a tour of the intake structure, the inspector noted that
routine surveillance testing continues to identify diesel fire
pump battery cell temperatures which are below specification (77
degrees F i 15 degrees F). This problem was previously identi-
fied during inspection 86-06.
Discussions with the licensee
indicated that the unit heater was verified to be operating
properly and that a review of the heater sizing was ongoing.
The inspector will review the results of this analysis in a
future inspection.
b.
Plant Maintenance and Outage Activities
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The inspector reviewed the action taken by the licensee in re-
sponse to IEN 86-106.
Boston Edison has experienced severe
erosion-corrosion (E-C) degradation of the extraction steam
piping, feedwater heaters and cascading drain piping at Pilgrim.
Starting-in 1980 the licensee implemented an extensive equipment
inspection and replacement program in these areas. This program
is ongoing. Two studies contracted by BEco in 1981 identified
six variables affecting the occurrence of E-C in steam piping.
Extensive ultrasonic examination of low pressure steam piping
conducted from 1980 to present identified significant degrada-
tion.
Four feedwater heaters and a large amount of piping was
replaced with less susceptible chrome-moly or stainless steel.
In response to the Surry event detailed in IEN 86-106, the
licensee expanded the scope of examinations to include the con-
densate and feedwater piping.
Inspections of fittings located
in accessible and high personnel traffic areas has begun with
completion expected during the current outage.
The criteria
established in response to the steam piping E-C problems, fluid
chemistry variables such as oxygen concentration, piping stress
levels, industry experience, and academic studies will be con-
sidered in establishing a prioritized inspection plan for the
remaining
piping.
Data
collection
from
the
inspections
currently underway will be used to refine the analysis method.
Licensee consideration is being given to extending this effort
to other systems including main steam and safety related piping
in the drywell .
Recommendations on operational and chemistry
practices, possible replacement material and design options are
under development.
The licensee appears to be aggressively
pursuing evaluation and
resoittion
of
this
industry-wide
concern.
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On January 16, 1987 the licensee reported via ENS that certain
electrical relays used in the core spray system injection valve
logic circuits are not seismically qualified. General Electric
type HGA relays may exhibit contact chatter when subjected to
seismic accelerations. A review by licensee engineering person-
nel identified an application of this relay type which could
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prevent automatic opening of the core spray injection valves
during the design basis earthquake. In excess of 200 HGA relays
exist at Pilgrim.
Licensee review of the application of these
relays is ongoing and will be evaluated by the inspectors under
existing item 50-293/86-37-09.
On January 23, 1987, miscommunication between site Quality Con-
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trol and Engineering sections regarding operability of a pipe
support led to an ENS report. Further evaluation concluded that
the pipe support was operable.
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As documented in inspection report No. 50-293/86-40, the A core
spray pump could not achieve rated flow during post maintenance
testing. The inspector reviewed the results of the licensee's
evaluation of the flow problem. During the investigation, the
licensee opened and inspected the pump discharge and test line
check valves (valves 1400-36A and 1400-35 respectively), removed
and inspected the test line restricting orifice, and performed a
boroscopic inspection of the core spray test line.
The only
problem identified was that the disc of valve 1400-35 was found
to be excessively loose on the hinged hanger. It was determined
that the obstruction caused by the deteriorated check valve
resulted in the reduced core spray flow.
After modifying the
internals.of check valve 1400-35, the A core spray pump satis-
factorily completed a full flow test on December 17, 1986.
c.
Surveillance Testing
On February 1,
1987, the inspector witnessed performance of
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Surveillance test 8.M.2-2.10.8.6, Diesel Generator B Initiation
by Loss of Offsite Power Logic. The as-run copy of the test was
Revision 5.
The procedure revision obtained by the inspector on
February 1, 1987 from the licensee's document control center
(DCC) was also revision 5.
During conduct of the test the
inspector noted several apparent technical deficiencies which
were referred to the licensee. The licensee later informed the
inspector that revision 6 of the procedure had been approved for
issuance on January 28, 1987. This revision appears to resolve
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the technical questions raised by the inspector.
However, it
appears the test run on February 1 had used an out-of-date pro-
cedure.
The licensee's Document Control Center issues revised
procedures first to the control roon.
DCC routinely refrains
from general issuance of the new revision until confirmation of
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receipt is received from the control room.
In this case the
delay resulted in the use of Revision
5,
four days after
Revision 6 had been approved.
The licensee stated that until
a long term solution can be achieved, control room confirmation
of procedure receipt would be hand carried to DCC. The licensee
subsequently reperformed the test using Revision 6.
The inspec-
tor will review licensee actions taken in this area.
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The inspector noted that no technical specifications for the
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reactor protection
system electrical
protection
assemblies
exist.
By letter dated September 24, 1980 NRC:NRR requested
that the lic,ensee provide a schedule for modification of the
reactor protection system (RPS) power supply and submission of
Technical Specifications addressing the modifications.
By let-
ter dated November 26, 1980, Boston Edison committed to instal-
latf or of GE-designed electrical protection assemblies and sub-
mission of the appropriate Technical Specifications.
NRC ap-
proval of the design was transmitted by letter dated July 28,
1982.
Submission of technical specifications was again reques-
ted. The EPAs, which function to provide RPS power supply trips
on overvoltage, undervoltage and underfrequency, were subse-
quently installed.
The licensee however, had not submitted
Technical Specifications nor revised their commitment.
In re-
sponse to the inspector's question the licensee submitted to
NRC:NRR, by letter dated February 17, 1987, a commitment revis-
ion stating that Technical Specifications are not warranted.
On February 11, 1987, the licensee reported via ENS that two of
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fo'ur main steam safety relief valves (SRV) had failed lift
pressure tests during checks in an offsite test facility. Tech-
nical Specifications require lift pressures of plus or minus one
percent of nominal setpoint.
One SRV was found to lift 1.4
percent high and the second 2.2 percent low.
Pilgrim Station
has four two-stage Target Rock SRV's. During the 1984 outage,
testing revealed significant failures with one SRV pilot poppet
found frozen on its seat.
In response to these failures the
licensee performed extensive metallurgical evaluations and re-
placed the pilot poppet seat material. While the lift pressures
of two of the four valves recently tested were slightly outside
the acceptable range, the results represent significant improve-
ment over past performance.
d.
Fire Protection
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On January 19, 1987, the licensee reported via ENS that the
diesel driven fire water pump had been declared administratively
inoperable. A nonconformance report (NCR) written on December
3,1986 identified a small tubing fitting in a pump seal water
line which did not meet QA inspection requirements.
The Boston
Edison Quality Assurance Manual (BEQAM) states that the exist-
ence of open NRCs requires that the affected system be declared
inoperable. This requirement was not recognized at the i.ime by
operations personnel who reviewed the NCR.
The equipment was
not declared administratively inoperable until January 19, 1987
when a deficiency report was issued by the QA department high-
lighting the situation.
The pump was functionally tested and
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was - found- to be capable of_ providing the required flow.
The
discrepancy was resolved and the pump returned to operable
status by January - 23, - 1987.
The redundant electric fire pump
was also inoperable during this period due to ongoing mainten-
ance.
The inspector questioned the licensee regarding the
possible existence of other unaddressed NCR's.
The licensee
stated that this incident appears to be an isolated occurrence.
The inspector had no further questions about this incident.
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On February 6,1987, the licensee reported via ENS discovery of
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several potentially degraded fire barriers. The licensee estab-
lished fire watches in the area in accordance with Technical
Specification requirements.
Similar barrier problems have been
reported several times during the outage as a result of ongoing
walkdowns.
Final walkdown results and licensee resolution of
findings will be reviewed under existing open item 50-293/
86-36-02.
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Boston Edison reported _on February 19, 1987 via - ENS that the
fixed dry chemical fire suppression system for the A Emergency
Diesel Generator (EDG) was found to have been inoperable since
December 23, 1986. Technical Specifications require that a con-
tinuous fire watch be established if this system is inoperable.
A maintenance request written on December 23, 1986 documented
that the pressure in one of three chemical storage bottles asso-
ciated with the system was found to be slightly below the mini-
mum required pressure.
At the time the MR was written, the
impact of the low pressure on system operability was not recog-
nized by operations personnel and, consequently, the required
compensatory measures were not taken. The dry chemical system
and the EDG were subsequently assumed operable during fuel move-
ment activities. Upon discovery of the problem on February 19
the licensee established a continuous fire watch in the area.
While no continuous fire watch was in place from December 23,
1986 to February 19, 1987 a roving fire patrol toured the area
on an hourly basis due to unrelated deficiencies.
Operation
personnel initiating the MR did not take appropriate action
regarding system operability.
The MR was mistakenly written
against the diesel generator system rather than the fire protec-
tion system. Because of this misclassification, the Fire Pro-
tection Group was not aware of the problem.
This situation
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appears similar to the case discussed above in which operations
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personnel did not recognize the impact of an NCR on diesel fire
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pump operability.
The sensitivity of the operations staff to
fire protection equipment operability and the ability of the
Fire Protection Group to monitor equipment status will be evalu-
ated during the next inspection period,
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The licensee's fire water system consists of two redundant fire
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water storage tanks (FWST) and two redundant fire water pumps.
On October 21, 1986 the diesel driven fire water pump was re-
moved from service. This pump was returned to operable status
on November 11, 1986.
On November 5,
1986 the B FWST was
drained for maintenance and remained unavailable until late in
November, 1986.
On November 11, 1986 the electric driven fire
water pump sustained serious damage while operating without a
suction source.
This incident was the subjet of special in-
spection 50-293/86-38.
The electric pump was not returned to
operable status until February 4,1987. This series of mainten-
ance activities and equipment failures resulted in a loss of
fire water system redundancy from October 21,
1986 until
February 4,1987. At all times during this period one of the
redundant FWSTs, one of the redundant fire water pumps or a
combination of equipr.;ent was inoperable.
Technical Specification 3.12.B requires that two operable fire
water pumps, two separate fire water supply tanks, and associ-
ated flow paths be maintained. With less than the above equip-
ment, the licensee must restore the inoperable equipment to
operable status within seven days or submit a report to the
Commission within thirty days outlining the procedures used to
provide for the loss of redundancy in this system. Contrary to
the above, as of January 23, 1987 fire water supply system re-
dundancy had been lost for three months.
No steps had been
taken by the licensee to compensate for this loss of redundancy
and no report had been submitted to the Commission describing
such plans. The inspector informed the licensee that the above
constituted
a
violation
of
the
Technical
Specifications
(87-03-01). When informed, the licensee aligned the onsite fire
truck with its suction - from the city fire water hydrant and
discharge into the fire main.
o.
Radiation Protection
On January 6,1987, the inspector noted that little corrective
-
action had been taken for a lost master high radiation area key.
On December 22, 1986, the Watch Engineer notified the Assistant
Chief Radiological Engineer that a master "R" key, number OP-5,
was missing from the control room. This key is normally used by
operations personnel and unlocks all plant areas with dose rates
between 1,000 mrems/hr and 10,000 mrems/hr.
In addition, the
key opens one of two locks on access doors to areas with radia-
tion doses greater than 10,000 mrems/hr.
The Watch Engineer
found the key ring in the control room at that time, but not the
key.
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The Assistant Chief Radiological Engineer initiated Radiological
Occurrence Report (R0R) 86-12-22-0993 which described the inci-
dent and immediate corrective actions on December 22, 1986, He
contacted the Chief Radiological Engineer (CRE) and the Radio-
logical Section Head at that time.
The following problems were noted during the inspector's review.
The immediate corrective action specified on the ROR was
--
not implemented.
The cl0R stated that the frequency of
surveillance checks of high radiation areas would be
increased. This was not done until January 3, 1987, fol-
lowing' the licensee review of a disabled high radiation
area lock'(ROR 87-01-03-003).
--
The licensee did not check the control room log to deter-
mine who had last returned the missing key to the _ control
room.
When the log was checked, the licensee determined
that the key was unaccounted for between December 19 and
December 22.
The licensee did not question all the operations staff who
--
had access to the key to determine if they may have inad-
vertently retained the key.
The licensee took no action to strengthen administrative
--
controls over keys issued from the control room following
the discovery of the missing key. At Pilgrim, high radia-
tion area access keys are issued at two locations, the main
health physics control point and the control room.
The
station key control procedures, 1.3.10 and 6.1-012, de-
scribe the health physics control point key issue method in
detail.
While the procedures mention the control room's
issue method, they do not discuss it in detail.
In response to previous incidents which occurred in the
last quarter of 1986 involving high radiation area control,
the Operations Section Manager issued detailed instructions
on key issuance in a memo on December 15, 1986.
These
instructions were not followed when the key was taken from
the control room between December 19 and 22. At the time
of the inspector's review, the licensee was not aware that
these instructions had been violated and subsequently found
that the memo had not gone to all operations personnel.
Subsequently,
the memo was
incorporated
into station
procedures.
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In addition, the licensee indicated that the key was pro-
bably . intentionally taken off the key ring. High radiation
area key rings have large disks attached to them, to pre-
vent an individual from inadvertently leaving the station
with a key in their pocket.
Removing the key from the key
ring eliminates this administrative control.
However, no
action was taken to address this problem until after the-
inspector's review.
--
Initially, neither the Operations nor the Radiological
Section Managers assumed the responsibility for determining
root cause and implementing corrective action for the lost
key.
Station procedure 1.3.10 indicated that the Chief
Radiological
Engineer in the Radiological Section was
responsible for the administrative control of high radia-
tion area keys as described in procedure 6.1-012. However,
procedure 6.1-012 assigned responsibility for the control
of the high radiation area keys in the control room to the
Chief Operating Engineer in the Operations Section. Subse-
quently,
the
Radiological
Section
Manager
and
the
Operations Section Manager jointly reviewed the incident.
All locked high radiation areas were subsequently reviewed to
determine if radiation levels were below 1,000 mrems/hr and the
areas could be unlocked.
In addition locked high radiation
areas were checked every four hours to ensure that they were
properly secured. New high radiation area locks were installed
after a high radiation area door was found open on January 24,
1987. The locks were changed because at the time the door was
found open no one should have had an authorized key to the area.
The licensee manned all locked high radiation areas continuously
between January 24 and the time the new locks were installed.
Also, the control room key issuance policy was changed to re-
quire operators to obtain high radiation area keys from the
health physics control point.
The keys in the control room
would only be used for emergencies by Operations personnel.
Technical Specification 6.13, "High Radiation Area" requires in
part that areas with radiation levels greater then 1,000 mrems/
hr be locked and the keys maintained under the administrative
control of the Shift Foreman or unit health physicist.
Failure
to maintain administrative control over master key (0P-5) to all
locked areas in the plant with radiation levels greater ti en
1,000 mrems/hr and failure to promptly take corrective action is
.
a violation (87-03-02).
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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'On February 18,1987. during a tour-of L the reactor building the
-
-
inspector observed two individuals working in the reactor __ vessel-
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instrument cages.- The area was posted:as contaminated and re-
quiring an RWP.
The applicab1_e ~ RWP required both -cotton and
rubber gloves. The individuals'were removing cables in prepara-
tion for. instrument replacement and were wearing only cotton
. gloves. The health physics technician responsible for the. area
stated that the individuals 'were - having difficulty performing
their task while wearing rubber gloves, and that he had allowed-
.the workers _ to remove _them after. having surveyed the area._ :No
.'
revision to the RWP. had .been processed.
The ongoing' work was -
not being performed in.-accordance with the existing RWP. The'
inspector _ informed licensee radiation protection department
management. The situation was of minor safety significance but
demonstrates a lack- of attention to detail and a . poor . example
~
,
being set by health- physics personnel.
Poor contamination con-
trol practices have been observed in this same area by inspec-
. tors _ several times- in the past.
In-. response the licensee sus-
pended the supervising technician for not enforcing the approved -
RWP.. The inspectors,will-continue to closely monitor this area.
'4.0 ~ Review of Plant Events
The inspectors:followed up on events occurring during the period to deter-
mine if licensee response was thorough and effective. . Independent reviews -
2
of the events were- conducted to verify the accuracy and completeness of
= licensee information.
a.
Licensee Management and' Organizational Changes
During
the
period
Boston
Edison
Company. named
Rear Admiral
' Ralph G. Bird, U. S. Navy (Retired) to head the nuclear organization.
Mr.. Bird, a 28 year veteran of the nuclear navy .and formerly a con-
sultant to the; nuclear power industry, will assume the position of
+
Senior Vice President-Nuclear.
'In' this capacity he will be respon-
sible for all aspects of the licensee's , nuclear operation. Bird will
report directly- to Stephen J.
Sweeney, Boston Edison's Chairman,.
-President and Chief Executive Officer.
A. Lee Oxsen, Vice President
for Nuclear Operations, and J.
Edward Howard, Vice President of.
!
Nuclear Engineering and Quality Assurance, will report directly to
Mr. Bird. The transition from James Lydon, Chief Operating Officer,
to_Mr. Bird was completed on February 20, 1987.
The~ licensee announced on February
4,
that Mr.
K.
Roberts, the'
current
Director
of
Outage
Management
at
Pilgrim,
replaced
'
~Mr. A. Pederson as the Plant Manager.
Mr. Roberts reports to the
Vice President of Nuclear Operations.
Mr. Pederson became a staff
assistant to the Senior Vice President-Nuclear. The site disciplines
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have been reorganized with training, fire protection, security, and
. emergency _ planning reporting directly. to the Vice: President of-
Nuclear Operations. Plant operations,, maintenance, radiation -protec-
tion, itechnical' . and . outage management groups report to the Plant
Manager. This realignment was effective' February 4, 1987.
.'b.
-Loss of Offsite Power Test
- The licensee conducted the once per cycle simulated loss of offsite
power : test on February 5,1987.
Execution.of the test was witnessed
by three NRC resident inspectors.
Independent review of test data
to confirm acceptable results _ was also performed by the inspectors.
During initial 1 test performance the licensee identified four. signif-
icant deficiencies: 1) the Emergency Diesel Generator Breaker closed
onto the A6 bus -later than design ~, 2) the secondary emergency power
source closed 'onto the A6 bus earlier than design, 3) the D RHR pump
-breaker closed earlier than design and 4) the auto transfer of 480-
VAC load center B6: failed to properly function. The first two defic-
iencies' resulted in the emergency diesel and secondary emergency
power supply breakers racing to _reenergize the bus. The-early start
of the D RHR pump resulted in the B and D RHR pumps starting in close
succession rather than the desired sequence. - The licensee initiated
-
Failure and Malfunction Reports (F&MR) for these three problems and
notified-_ the NRC via ENS of the failures. .The applicable timing
relays were ~ recalibrated, and the test reperformed with acceptable
results.
Licensee followup of these four test discrepancies will be
reviewed during a future inspection (87-03-03).
During the test, it was noted that the power source _ transfer scheme
'for 480 VAC- vital bus 86 failed to function.
Bus B6 supplies AC
- power to vital loads such as both RHR system injection valves,1 and
containment isolation valves.
The -inspector noted that the opera-
bility of the B6 bus transfer mechanism is tested only once per oper--
ating cycle.= Although the operability.of the transfer scheme for Bus
B6 is not included as a test acceptance criterion, it is a failure of
,
p
a safeguards. design feature which could adversely impact plant
'
safety,
i -
A Failure and Malfunction Report (F&MR) concerning the bus B6 trans-
i-
fer failure was not initiated until February 17, 1987 when the in-
spector questioned the documentation and followup of the failure.
Station
Procedure No.
1.3.24,
Failure and Malfunction Reports,
Revision 14, indicates that the purpose of the F&MR is in part to
i:
ensure that an initial internal review and safety assessment is made
.
of events of potential safety significance. The reports are submit-
,
ted-to the Nuclear Watch Engineer who reviews them to identify events
.
that are related to Technical Specification requirements and that may
.
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require NRC notification.
Issuance of a F&MR also ensures perform-
ance of a root cause analysis.
Procedure No.1.3.24 states that an
F&MR shall be initiated whenever failures or malfunctions are iden-
tified during surveillance testing of safety related components that
do or could prevent systems from fulfilling their individual func-
tions. The procedure also states that the individual responsible for
initiation of the report shall be the person who first identifies an
abnormality or a deviation from normal condition.
Failure to initi-
ate an F&MR following the discovery of the B6 bus transfer failure en
February 5,1987 is a violation of Procedure No. 1.3.24 (87-03-04).
Several-previous items of noncompliance have been issued for licensee
failure to properly utilize the corrective action program.
c.
Reactor Defueling Preparations and Conduct
The licensee conducted reactor defueling between February 6 and
February 13, 1987. All 680 fuel assemblies were transferred from the
reactor vessel to the spent fuel pool. The inspectors reviewed the
defueling activities periodically.
The review included preparations
and prerequisites for core alterations, witnessing of fuel movement
from the refueling floor, and control room observations.
The inspector attended the onsite Operations Review Committee (ORC)
meeting (#0RC 87-20) held on February 6,1987 to observe the conduct
of the pre-core-offload checklist review.
The committee reviewed
surveillance test records, maintenance requests, and Failure and
Malfunction Reports to verify operability of vital systems including
the emergency diesel generators, the core spray system, the standby
gas treatment system, refueling floor ventilation process radiation
monitors, refueling platform and reactor manual control
system.
The inspector reviewed the pre-refueling checklist (OPER 10). This
checklist was completed prior to fuel movement and included verifi-
cation that the Onsite Review Committee had granted permission to
move fuel. A review of control room logs indicated that the periodic
testing of the refueling equipment (OPER 13, OPER 14) had been per-
formed. Procedure No. 4.3, Fuel Handling, was reviewed to ensure it
adequately addressed Technical Specification (TS) requirements for
fuel handling.
The inspector also observed on-the-job training of
two licensed operators on the new refuel bridge. The training was
performed in accordance with a prepared training check list and
appeared adequate.
Fuel movement activities were observed from the
refueling floor and from the control room. A Senior Reactor Operator
.
supervised the refueling bridge activities and was assisted by a
licensed Reactor Operator and a reactor engineer.
The activities
were well controlled and the communication between the refueling
bridge and the control room were adequate.
.
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16
The inspector reviewed the status of the standby gas treatment (SBGT)
-
system in preparation for refueling operations.
The licensee sub-
mitted Licensee Event Report (LER) No. 86-21, one LER supplement and
two 10 CFR Part 21 reports which described various potential compon-
ent failures within the SBGT system that could result in offsite
radiation doses exceeding 10 CFR 100.11 limits during a design basis
loss of coolant or fuel handling accident. The licensee determined
that each of the postulated SBGT. system failures could result in
reducing the efficiency of the SBGT system charcoal filters.
These
filters are installed to reduce the potential release of radioactive
iodine to the environment during the design basis accidents noted
above.
To allow refueling during the current outage without. modifying the
SBGT system, the !!censee performed safety evaluation No. 2027 which
addressed the operability of the system under current plant condi-
tions. The inspector discussed this safety evaluation with engineer-
ing, operations and technical support personnel and compared it to
the requirements of the Technical Specifications and the commitments
made in the Final Safety Analysis Report.
The safety evaluation
indicated that, for a fuel handling accident after October 4,
1986-
(175 days after the April 1986 shutdown), and assuming a zero percent
charcoal filter efficiency, the offsite doses would be five orders of
magnitude less severe than for the design basis event.
This was
because the radionuclides produced during power operation which would
be absorbed by the charcoal have been allowed to decay.
The safety
evaluation was reviewed and approved by the Operations Review Commit-
tee during meeting no.86-146.
The inspector also noted that to further mitigate the consequences
of a fuel handling accident, the licensee initiated Temporary Modif-
ication No. 87-02 to maintain the deluge system isolated from the
SBGT system and to cause the crosstie damper to permanently remain
open.
In addition, the licensee modified the SBGT operating proced-
ure to require that a motor operated discharge damper be closed in
the event that one of the SBGT fans is stopped. This addresses the
backdraft damper failure issue described earlier in this inspection
report. The inspector had no further questions regarding SBGT system
operability for refueling. The licensee is evaluating modifications
to the system which must be installed prior to startup from the cur-
rent outage. These modifications will be reviewed by the NRC during
a future inspection and are the covered under existing open item
50-293/86-29-03.
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5.0 -Review of LER's
,
' ' LER's 1 submitted ;to NRC:RI were reviewed to verify that .the details were-
- clearly reported, -including ' accuracy of the description of cause and ade-
.
quacy. of corrective action.
The -inspector. determined -whether further
- information 'was . required from the' licensee, whetherigereric> implications
-were ; indicated, and whether - the event. warranted 'onsite' followup.
The
following LER's were reviewed:-
LER No.
Event Date'
Report Date
Subject
86-26-00
11/04/86
12/04/86
Failure to perform--Sur-
veillance Tests of Stand-
by. Gas Treatment (System
and Liquid -Radioactive
Effluent Flow - Monitor.
86-28-00.
12/22/86
01/21/87
Failure
to
- Recognize
-
Effects
of
Electrical
,
Isolation Resulting' in
an Unplanned 'ESF Actua-
tion.
86-29-00'
12/23/86
01/22/87
Loss of Offsite: Power
While Removing Salt from
Insulators.
LER 86-26-00 describes the licensee'.s failure to perform adequate surveil-
~
lance testing of the standby gas treatment system- (SBGT) -and the radio-
active-liquid effluent flow rate device.
The LER states that .a technical
review of station surveillance requirements and procedures had:been con-
' ducted but that- Amendment 89 was not included in this review because of '
its late issuance.
Inadequate administrative controls resulted in the
failure to properly' implement .the new requirements incorporated by .this
amendment. .The inspector pointed out that the SBGT system-test require-
ments were .not altered by Amendme'nt 89 and had existed' during the refer-
enced technical review. The root cause of this deficiency:does not appear
to be'related to technical specification amendment issuance. The LER also
stated that there was no evidence of the failure of the radioactive ~efflu-
ent flow monitor to perform its design function.
In fact, this device is
not and has not been functional due to a design deficiency. The Opera-
!
tions Department maintains a procedure to calculate discharge flow rate
- based on tank levels.
The licensee stated at the ' exit meeting that an
,
updated LER would be issued.
j
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The . inspector. noted .that the root cause for the spurious ESF actuation . in
-
LER 86-28 was _ attributed to operator error. L However, a lack of preplann-
.ing prior 'to the isolation' of electrical . bus - B-18.was also a major con-
. tributor _to the event. :At the exit meeting,_ the licensee agreed and
stated. that .the ' outage- manual had been changed to - require. more planning
prior ' to major electrical isolations. This event ~ was _ reviewed in NRC
~
-
inspection 50-293/86-40. Thc sequence of events in the LER is not fully
consistent with' that . in inspection report' 86-40.
The inspector-checked
.the LER sequence against plant logs and drawings and concluded that the
LER was-correct. The licensee plans to update the LER to more accurately
reflect the root cause and corrective ' action' for the isolation problem.
The inspector noted that corrective actions were not completely.. described
in LER.86-29. Specifically, the licensee required that the control room-
~
be notifi_ed prior to each switchyard washing sequence. This assured that
the operations staff was ; aware of the washing and could prepare for a
potential loss of -offsite power.
Also, the licensee is -considering an
insulator design: change to lessen the need to wash the switchyard. This
incident was also. reviewed -in NRC report 50-293/86-40.~:The licensee also
plans to update this LER to more completely describe the corrective ac-
tions taken.
6.0 Management Meetings
At periodic intervals during the course of_the inspection period, meetings
were held with senior facility management to discuss the inspection scope
and preliminary findings of the resident inspectors. No written material
was given to the-licensee that was.not pre'vious'y available to the public.
On January 20,1987, _the licensee met with NRC management in the Region I
offices to discuss recent problems .in'the area of.the station fire protec-
tion program. Minutes of this meeting are documented in report 50-293/
t
- 87-07.
A meeting between NRC Region I and Boston _ Edison senior management was
held at the licensee's Chiltonville training center on February 2,1987.
This -is one of a series of ongoing management meetings which provide NRC
management the opportunity to closely monitor the progress of licensee
' improvement programs.
Minutes ' of this meeting are documented in report
50-293/87-08.
.
______________..___-_________.____.____m_
- _ _
r=
,
..
. - e, s
Attachment I to Inspection Report 50-293/87-03
~ Persons Contacted
- R. Bird, Senior Vice President, Nuclear
L. Oxsen, Vice President, Nuclear 0perations
K. Roberts, Nuclear Operations Manager
A. Pederson, Staff Assistant - Senior VPN
D. Swanson, Nuclear Engineering Department Manager
R. Fairbank, NED, Licensing and Analysis Section Head
N. Brosee, Maintenance Section Head
T. Sowdon, Radiological Section Head
J. Seery, Technical Section Head
P. Mastrangelo, Chief Operating Engineer
R. Sherry, Chief Maintenance Engineer
N. Gannon, Chief Radiological Engineer
F. Wozniak, Fire Protection Group Leader
C. Higgins, Security Group Leader
- Senior licensee representative present at the exit meeting.
.